ML20062E302

Safety Evaluation Supporting Util Analysis of Main Steam Line Break W/Continued Feedwater Addition

ML20062E302
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/28/1982
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20062E301	List: ML20062E298 ML20062E302
References
IEB-80-04, IEB-80-4, NUDOCS 8208090289
Download: ML20062E302 (7)

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SAFETY EVALUATION REPORT

• 1 MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION

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ARKANSAS NUCLEAR ONE-UNIT 2 Docket No. 50-368 1.0

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Introduction.

In the summer of 1979, a pressurized water reactor (PWR) Licensee submitted a report to the NRC that identified a deficiency in its original analysis of the containment pressurization resulting from

, ~a postulated main steam Line break (MSLB). A reanalysis of the c o'nt a i n me nt pressure response fotLowing a MSLB was pe rf ormed, and it was determined that, if the auxiliary f eedwater (AFW) system continued to supply feedwater at runoat c ondi tio ns to the steam generator that had experie nced the steam Line break, the containment design pressure would be exceeded in approximately 10 minutes. In ot he r words, the long-term b Lowdown of the water supplied by the AFW system had not

• been conside red in the earlier analysis. -

l On O ctobe r 1,1979, the foregoing information was provided to aLL ho lde rs of operating Licenses and construction permits in IE i

l Information Notice 79-24 C23. Another licensee perf ormed an l

ac ci dent ar.alys i s revi ew pursua nt to the inf ormation f urnished in l

l the above cited notice and discovered that, with of f site electrical power available, the condensate pumps would f eed the affected steam generator at an excessive rate. This exce.ssive feed had not been considered in its analysis of the postulated MSLB a c ci de nt.

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. A third Licensee inf ormed the NRC of an error in the MSLB analysis ' "7' f or their plant. For a zero or low power condition at the,en,d of core Life, the Licensee identified an incorrect postulation that the '

startup feedwater control valves would remain positioned "as is" during the transient. In reality, the startup feedwater control valves wilL ramp to 80% f ulL open due to an override signal resulting from the low steam generator pressure reactor trip signal. .

Reanalysis of the events showed that the rate of feedwater addition to the af fected steam generator associated with the opening of the startup valve would cause a rapid reactor cooldown and resultant reactor-return-to power response, a condition which is beyond the plant's design basis.

FolLowing the identification of these deficiencies in the original MSLB ac ci de nt analysis, the NRC issued IE ButLetin 80-04 on F eb rua ry 8, 1980. This.butLetin required alL licensees of PWRs and ce rt ain nea r-t e rm PWR operating Li c en's e appli cant s to do the following:

"1. Review the containment pressure response analysis to determine if the potential for containment overpressure for MSLB 'inside

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containment included the impact of runout flow from the auxiliary feedwater system and the i mpact of other energy

. sources such as continuation of feedwater or condensate flow.

l In your review, consider your ability to detect and i solate the damaged steam generator f rom these sources and the ability of the pumps to remain operable after extended operation at l '

runout flow.

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2. Review your analysis of the reactivity increase which results ..

from a MSLB inside or outside containment. This review.should. .

consider the reactor cooldown rate and the potential f or the .

reactor to return to power with the most reactive control' rod in the fully withdrawn position. If your previous analysis did not consider alL potential water sources (such as those Listed in 1 above) and if the reactivity increase is greater than previous analysis indicated, the rep o rt of this review should include:

a. The bounda ry conditions for the analysis, e.g., the end of Life shutdown margin, the moderator temperature coefficient, power level and the net effect of the associated steam generator water inventory on the reactor sys tem cooling, etc.;

b. The most restrictive single active failure in the safety injection system and the effect of that failure on delaying the delivery of high concentration boric acid .

solution t'o ,the reactor coolant system; .

c. The effect of extended water supply to the affected steam generator on.the core criticality and return to power; and i

d. The hot channel f actors corresponding to the most reactive rod in the fully withdrawn positions at the end of Life, and the Minimum Departure f rom Nucleate Boiling Ratio (MDNBR) values for the analyzed transient.

3. If the potential f or containment overpressure exists or the reactor retu rn-to powe r response wo rs ens, p rovide a proposed

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corrective action and a schedule for completion of the ~

corrective action. If the unit is operating, provide',a . .

, description of any interim action that wiLL be taken until l

the proposed c'orrective action is completed."

i i Following the Licensee's initial response to IE Dulletin 80-04, a f .

[ request f or additional inf ormation was developed to obtain all the information necessary to evaluate the Li,censee's analysis.

The results of our evaluation for Arkansas N u c l e a r. O n~e Uni,t 2 i

(ANO -2) are- providedibelow. .I - -

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3 j 2.0 . Evaluation f Our consultant, the Franklin Research Center (FRC), has reviewed the submittals made by the Licensee in response to IE Bulletin i

, 80-04, and prepared the attached Technical Evaluation Report. We have reviewed this evaluation and concur in its bases and findings.

3.0 ~ Conc ~l'dsion _ .

Based on our review of the enclosed Technical Evaluation Report, the following conclusions are made regarding the postulated MSLB with continued feedwater addition for Maine Yankee:

l 1.

There is no potential for containment overpressurization resulting from a MSLB with continued feedwater addition because the main feedwater system is isolated and auxiliary

• feedester a ct uat ion systes prevents the affected steam generator from being fed; 4 .

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?> l The electrical instrumentation and controls needed to detect 2.

and initiate isolation of the affected steam generator a.nd . .

f eedwater sources meet the safety grade requirements of IEEE Standard 279-1971;

3. The AFW pumps wiLL not ex pe ri enc e runout conditions; therefore, they wiLL be able to carry out their intended function without incurring damage during a MSLB;

4. AlL potential water sources were previously identified; therefore, the FSAR reactivity increase analysis remains valid.

5. No f urther action is required by the licensee rega rdi ng IE Bulle tin 80-04 4.0 References

1. IE Bu t Le tin 80-04, -%alysis of a -PWR Main-1 team- Li ne B reak -

with Continued Feedwater Addition," MRC Of fice of Inspection and Enforcement, Feb rua ry 8, 1980

2. IE Inf o rmation Not ice 79-24, "Ove rp r es sur 'za tion of the Containment of' a PWR P la nt After a Main Steam ,Line Break l,"

NRC Office of Inspection and Enf orcement, O ctobe r 1,1979

3. D. C. Trimble (AP8L), Letter to K. V. Seyfrit (NRC, Region IV)

Subject:

ANO-1 and ANO-2 Response to IE ButLetin No. 80-04 May 27, 1980 4 Arkansas Nuclear One - Unit 2, Final Saf ety Analysis Report, through Amedment No. 46, Arkansas Power & Light Company,

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May 19, 1978

5. Technical Evaluation Report, T ER-C5 506-119, "PWR Main Steam Line Break with Continued Feedwater Addition - Review of

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A cceptance Criteria," F rank li n Resear ch Cent er, Novembe r 17, 1981 .

6. IEEE Standard 279-1971, "C ri te ri a f o r P rot ect ion Sys t ems for

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Nuclear Power Generating Stations," Institute of Electrical and Electronics Engineers, New York, NY,1971 7 . NUREG-0800, " Standard Review P la n, S ect ion 15.1.5, "S t eam System Piping Failures Inside and Outside of Containment (PWR)", NRC, J u ly 19 81 8 . ANS / ANSI-4.5-1980, "C ri te ria f or Ac cident Monito ring Functions in Light-Wate r-Cooled Reacto rs," Ame ri can Nucle a r Soci ety, Hinsdale, IL, December 1980

9. Regulato ry Guide 1.97, " Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Ac cident," Revision 2, NRC, Decembe r 1980 -

10. ANS-51.7/N65'8-1976, " Single Failure Criteria f or PWR Fluid Sys tems," Ame ri can Nucle ar Soci ety, Hi nsdale, IL, June', 1976

11. R egu la to ry Gui de.1.26, "Quali ty G roup C la s si fica tions and Standards f o r W at e r,' S t ea m, and Radioactive-Waste-Containing Components of Nuclear Power Plant," Revision 3, NRC, Februa ry 1976 ,

12. NUR EG-0588, " Int erim St af f Pos i tion on Envi ronment al

-Qualification of Safety-Related Electrical Equipment,"

Revision 1, NRC, J u ly 1981

13. IEEE Standard 338-1971, "IEEE Trail-Use C ri t e ri a f or the ,

Periodic Testing of Nuclear Power Generating Station i

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}jii Systems," Institute of Electrical and Electronics Engineers, -

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14 Regulato ry Guide 1.22 " Periodic Testing of Protection System Actuation Funct ions," NRC, February 1972 Branch Technical Position'(BTP) ASB 10-1,

15. "D es i;n Guidelines for Auxiliary Feedwater System Pump Drive and Power Supply D i ve rs i ty f or P res surized Water Reactor Pla nts," Revi sion 2, NRC, J u ly 19 81 ,

httachment:

FRC Technical Evaluation Report 6

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